ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 17

Document Title:

Atomics International – 

Occupational External Dose

Document Number: ORAUT-TKBS-0038-6 

Revision: 01

Effective Date: 11/16/2006

Type of Document: TBD

Supersedes: Revision 00 

Subject Expert: Leo G. Faust

Approval: Signature on File Approval Date: 11/14/2006

Melton H. Chew, Document Owner

Approval: Signature on File Approval Date: 11/15/2006

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 11/15/2006

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 11/15/2006

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 11/16/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 2 of 17

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

02/22/2006 00 New technical basis document for the Energy Technology 

Engineering Center – Occupational External Dose. First approved 

issue. Initiated by Melton H. Chew.

11/16/2006 01 Approved Revision 01 to change the site name to Atomics 

International throughout the document in keeping with DOE site 

naming conventions. Updated required language in Section 6.1. 

Constitutes a total rewrite of document. Incorporates internal formal 

review comments and corrects typographical errors. Adds a 

paragraph to section 6.1.2 and clarification to missed dose and 

magnitude of background dose, Sections 6.2 and 6.8 respectively, as

requested by DOL. This revision results in no change to the 

assigned dose, and no PER is required. Training required: As 

determined by the Task Manager. Initiated by Melton H. Chew.
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 3 of 17

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

4

6.1 Introduction ..............................................................................................................................5

6.1.1 Purpose........................................................................................................................6

6.1.2 Scope...........................................................................................................................6

6.2 Missed Dose ............................................................................................................................6

6.3 Radiation Energies and Percentages at Selected Facilities...................................................... 7

6.4 Neutron Radiations and Percentages.......................................................................................8

6.5 Recorded Dose Practices....................................................................................................... 10

6.6 Interpretation of Reported Doses ........................................................................................... 11

6.7 Adjustments to Recorded Dose.............................................................................................. 11

6.8 Bias and Uncertainty..............................................................................................................12

References ........................................................................................................................................13

Glossary ............................................................................................................................................15

LIST OF TABLES

TABLE TITLE

6-1 Estimated annual missed photon, beta, and neutron dose .......................................................

PAGE

7

6-2 Beta and photon energies and percentages at reactors and critical facilities............................ 8

6-3 Beta and photon energies and percentages at support facilities............................................... 9

6-4 Facilities, neutron energies, percentages, and correction factors ........................................... 10

6-5 Recorded dose practices........................................................................................................ 10

6-6 Interpretation of reported dose ............................................................................................... 11

6-7 Adjustments to recorded dose................................................................................................ 11

6-8 Bias and uncertainty...............................................................................................................12
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 4 of 17

ACRONYMS AND ABBREVIATIONS

AI Atomics International

D deuterium

DOE U.S. Department of Energy

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

hr hour

keV kiloelectron-volt, 1,000 electron-volts

MDL minimum detectable level

MeV megaelectron-volt, 1 million electron volts

mR milliroentgen

mrad millirad

mrem millirem

MWt megawatt (thermal) 

NIOSH National Institute for Occupational Safety and Health

NTA nuclear track emulsion, type A

NVLAP National Voluntary Laboratory Accreditation Program

OW open window

PIC pocket ionization chamber

POC probability of causation

QF quality factor

SNAP Systems for Nuclear Auxiliary Power

SRE Sodium Reactor Experiment

T tritium

TBD technical basis document

U.S.C. United States Code

WB whole body

wk week

§ section or sections
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 5 of 17

6.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions for particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 

(42 U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures 

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 6 of 17

6.1.1

The purpose of this TBD is to describe external dosimetry systems and practices at Atomics 

International (AI). This document discusses historical and current practices at the AI facilities that, 

unless otherwise noted, used the same dosimeters and external dosimetry programs (Rutherford and 

Barnes 2006) in relation to the evaluation of external exposure data for monitored and unmonitored 

workers. 

Purpose

6.1.2

This TBD contains supporting documentation to assist in the evaluation of occupational external 

doses from processes that occurred at AI. An objective of this document is to provide supporting 

technical data to evaluate, with assumptions favorable to claimants, occupational external doses that 

can reasonably be associated with worker radiation exposures. This document addresses the 

evaluation of unmonitored and monitored worker exposure, missed dose, and the bias and uncertainty 

associated with the monitoring of external dose.

Scope

The AI facilities, which include four locations, have been identified in various ways over time. This 

TBD uses AI to refer to all locations unless more specific location information is warranted. In that 

context, AI includes Area IV of the Santa Susana Field Laboratory (which has also been known as the 

Nuclear Development Field Laboratory, the Liquid Metal Engineering Center, and the Energy 

Technology Engineering Center), portions of the Downey facility, the Vanowen Building at the Canoga 

facility, and the De Soto facility.

North American Aviation (NAA) entered into a contract with AEC to conduct nuclear research 

operations at Area IV and the Downey, Canoga and De Soto sites. At that time, Atomics 

International, an internal division of NAA, was the company’s designated nuclear research and 

development division. In addition to the employees of the AI division, other employees of NAA who 

worked at any of the above sites during the AEC contract period are potentially eligible for EEOICPA 

benefits.” (DOL)

AI had its own dosimeter in the beginning that was very similar to the multielement film dosimeters 

used at the Y-12 plant (ORAUT 2006a) and the Hanford Site (ORAUT 2004) based on the design 

developed by Pardue, Goldstein, and Wollan (1944) as well as pocket ionization chambers (PICs or 

pencil dosimeters). Both penetrating and nonpenetrating doses were recorded. AI started using 

commercial vendors in the early 1960s and continued that practice throughout the rest of its operating 

life. 

6.2 MISSED DOSE

Missed doses at AI facilities generally resulted from dosimeter minimum detectable levels (MDLs) and 

exchange periods. In reviewing an individual dose record from December 31, 1956, to December 30, 

1957, entries were both weekly and biweekly with the former starting in May 1957 (Author unknown 

1956). This would suggest that exchange periods varied with time, job, and individual. AI (1960) 

states, “Normally film badges are to be worn for monthly periods except where the possibility of 

exceeding 100 mrem/wk of exposure is expected. Then the badges are to be analyzed more 

frequently.” Therefore, missed doses before 1963, if not derived by review of individual records, 

would be based on weekly exchanges and would be applicable to both Hp(0.07) and Hp(10). Starting 

in 1963, most dosimeters were processed by their vendor (Landauer or its predecessor) (Garcia and 

Carpenter 1963a). On occasion, workers wore special dosimeters to monitor nonroutine work such as 

“hot jobs or cell entries” and always without their regular dosimeter. These special dosimeters were 

always worn in pairs; the results were averaged, and that average result was sent to the vendor for 
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 7 of 17

inclusion in the total dose for that period’s regular dosimeter result. It has not been determined if the 

special dosimeter procedure was in effect before 1963. Extremity finger ring film dosimeters were 

worn on both hands, and individual results were recorded for each hand. It has not been determined 

when this practice started, but it was in place in the December 31, 1956, dose record (Author 

unknown 1956) and included what appears to be the same exchange periods.

In cases of lost or destroyed dosimeters, results were derived from past results of similar work, 

coworker results, or the product of instrument measurements and time spent in the radiation zone. 

These practices are typical of sites with similar circumstances.

Neutron doses were measured with neutron track analysis, type A (NTA) film beginning with the start 

of reactor operations and the use of Van de Graaff accelerators. Both fast and thermal neutrons were 

measured and recorded as whole-body (WB) dose in rem. The NTA film is not effective at energies 

<0.5 MeV or at exposures of <50 mrem (ORAUT 2006a). It has not been determined what quality 

factors (QFs) were used. However, because a Po-Be neutron source was used for calibration, a 

reasonable assumption would be QF = 10 (Garcia 1970).

Missed doses for unmonitored employees could be as much as 500 mrem or 10% of whatever 

standard was in effect at the time of employment as allowed by AEC and it’s successors (AEC Manual 

Chapter 0524 and DOE Order 5480.11). However, it is very unlikely that anyone could have been 

exposed to this level since everyone who entered a controlled area was issued a dosimeter (AI, 

1960). The unmonitored workers may have been exposed to annual doses, above background, 

ranging from 10 to 40 mrem (ORAU-TKBD 0038-4, 2006) if they worked in Area IV. 

Table 6-1 lists the periods of use, types of dosimeters, exchange periods, MDLs, and estimated 

annual missed doses. 

Table 6-1. Estimated annual missed photon, beta, and neutron dose (rem). 

Perioda Dosimeterb Exchange period MDLc Estimated annualmissed dosed

1954–1962 Pocket dosimeter (PIC)

Site-specific two-element film

Daily 0.005 0.625

Weekly 0.04 1.04

Biweekly 0.04 0.52

Monthly 0.04 0.24

1963–1979 Landauer multielement film Monthly 0.04 0.24

Quarterly 0.04 0.08

1980e–present Landauer multielement film Quarterly 0.01f 0.02

1954–present NTA film Bi-weekly <0.05 0.650

Monthly <0.05 0.300

a. Estimated use periods for first entry. Landauer or its predecessor was the vendor in 1963 and maybe 

earlier (Garcia and Carpenter 1963a).

b. Neither the actual number of dosimeter elements of the first entry dosimeter nor the period of use of PICs 

has been determined. 

c. Estimated MDLs for each dosimeter in the workplace even though many doses were reported at less 

than the MDL.

d. Estimated annual missed dose calculated using MDL/2 from NIOSH (2002).

e. Neutron dosimeters continued monthly exchanges (Tuttle 1979).

f. Yoder (2005). 

6.3 RADIATION ENERGIES AND PERCENTAGES AT SELECTED FACILITIES

As described in ORAUT (2006b), there were many different types of facilities and processes at AI and 

other facilities during their periods of operation; these include reactors, critical test facilities, fuel 

preparation and postirradiation examination facilities, accelerator and calibration facilities, and support 
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 8 of 17

facilities. Most reactors were low power (a few kilowatts), with the maximum being 20 MWt, and all 

had relatively short operating histories. The major accelerator was a Van de Graaff ,DeuteriumTritium) machine producing neutrons with a maximum energy of 14 MeV. The fuel examination and 

manufacturing facilities, reactors, and critical facilities handled fissionable fuels with various 

enrichments, mostly compounds of uranium including carbides. They also handled relatively small 

quantities of plutonium and thorium with the exception of Buildings 4023, 4029, 4030, and 4363. AI

did not perform separations of irradiated fuel and, therefore, there were minimal gross fission product 

problems. AI did declad some fuels in the hot cells, which resulted in considerable quantities of those

fission products with high fission yields (e.g., 90Sr and 137Cs). There was some 152Eu and 154Eu along 

with some 3

H. Tables 6-2 and 6-3 list the facilities and related data.

Table 6-2. Beta and photon energies and percentages at reactors and critical facilities. 

Process/

building Descriptiona

Operations Radiation

type

Energy

Begin End (keV) Percentage

All Reactors 1956 1980 Beta >15 100

Photons 30–250 25

>250 75

4143 SRE reactor 1957 1964 Beta >15 100

Photons 30–250 25

>250 75

4010 SNAP Experimental Reactor and 

SNAP 8 Development ReactorS8ER

1959 1965 Beta >15 100

Photons 30–250 25

>250 75

4024 SNAP 2 Development ReactorS2DR 1961 1962 Beta >15 100

SNAP 10 Flight Simulation 

ReactorS10FS3

1965 1966 Photons 30–250 25

SNAP test facilities 1971 1971 >250 75

4028 Shield Test Reactor and

Shield Test and Irradiation 

ReactorSTIR

1961 1972 Beta >15 100

Photons 30–250 25

>250 75

4059 SNAP 8 Development ReactorS8DR 1968 1969 Beta >15 100

Photons 30–250 25

>250 75

4009 Organic Moderated ReactorOMR, and 

Sodium Graphite ReactorSGR

1958 1967 Beta >15 100

Photons 30–250 25

>250 75

4100 Advanced Epithermal Thorium 

ReactorAETR

1960 1974 Beta >15 100

Photons 

(thorium)

30–250 25

>250 75

a. SNAP = Systems for Nuclear Auxiliary Power; SRE = Sodium Reactor Experiment. 

6.4 NEUTRON RADIATIONS AND PERCENTAGES 

Table 6-4 lists facilities with neutron radiations. They are reactors, accelerators, and fuel storage 

facilities. The table includes Buildings 4005 and 4064, the plutonium fuel storage facilities, buildings 

4005, which were used for Pu storage, from 1967- to 1987 and 4064 from 1958- to 1993, respectively. 

However, these facilities are assumed to be a negligible contributor to neutron doses due to the 

limited quantities of fuel present at any one time. As near as can be determined, NTA film was 

incorporated in the film dosimeters if there was a potential of exposure >100 mrem in those facilities 

where neutron exposures were possible. How that was determined has not been found. In the dose 

record, there are entries in the “n” column. It is assumed that the dose recorded was the result of fast 

neutron exposure. 
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 9 of 17

A neutron survey report of the SRE Hot Cell containing a 14-MeV neutron generator (Clow 1966) lists 

dose rates of 75 mrem/hr fast and 11.8 mrem/hr thermal with “the assumption that all fast neutrons 

are 14 MeV.” Clow also states, “As further surveys are taken and the spectrum completely analyzed, 

the dose rates may be reduced.” These surveys were done on the roof of the facility and were higher 

than those taken at “a window” (62.2 mrem/hr fast and 7.1 mrem/hr thermal) or ”at the console” 

(7.1 mrem/hr fast and 2.2 mrem/hr thermal). The Van de Graaff accelerator with its D-T reaction 

generates 14-MeV neutrons that are quite monoenergetic in the 0° and 180° directions (Etherington 

1958), which might support the assumption that all fast neutrons can be treated as 14 MeV. 

Table 6-3. Beta and photon energies and percentages at support facilities. 

Building Description

Operations

Radiation

Energy 

Begin End (keV) Percentages

*SRE-4003, 4163, 

4041, 4654, 4689, 

4653, 4606, 4773

Support facilities 1954 1964 Beta >15 100

Photons 30–250 25

>250 75

*4020 Hot Laboratory 1957 1988 Beta >15 100

Photons 30–250 25

>250 75

*4064 Fuel Storage 1958 1993 Beta >15 100

Photons 30–250 25

>250 75

4011 Radiation Instrument 

Calibration Laboratory

1984 1996 Beta >15 100

Photons 30–250 25

>250 75

*4021/4022 Radiation Materials 

Handling Facility

1959 Present Beta >15 100

Photons 30–250 25

>250 75

4100 Calibration Laboratory 1985 Present Beta >15 100

Photons 30–250 25

>250 75

4363 Mechanical Component 1956 1963 Beta >15 100

Photons 30–250 50

>250 50

4029 Radioactive Measurement

Laboratory

1959 1974 Beta >15 100

Photons 30–250 25

>250 75

4030 Van de Graaff accelerator 1960 1964 Beta >15 100

Photons 30–250 25

>250 75

4023 Liquid Metal Component 

Testing

1962 1986 Beta >15 100

Photons 30–250 50

>250 50

*4005 Union Carbide Fuel Pilot 

Plant

1964 1967 Beta >15 100

Photons 30–250 25

>250 75

*4055 Nuclear Materials 

Development Facility

1967 1979 Beta >15 100

Photons 30–250 25

>250 75

*For operations involving Th-232, 1964 and 1979. 
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 10 of 17

 

Table 6-4. Facilities, neutron energies, percentages, and correction factors. 

Facility Source

Neutron energy 

(MeV)

Default dose % and 

correction factors

Reactors Reactors 0.1–2.0, WrR=20 100%

Pu fuel storage Correction factor = 1.91

Buildings. 4005 & 4064

Spontaneous fission and 

alpha- neutron reactions

4030 Van de Graaff

accelerator

D-T reaction 2–14 max, wR=20 100%, correction factor = 1.32

The distribution of energies and International Commission on Radiological Protection Publication 60 

conversion correction factors (ICRP 1991 and ORAUT 2006c) are also listed in Table 6-4. The 

correction factor for the 2- to 14-MeV energy group was calculated from data in the Y-12 TBD 

(ORAUT 2006a).

6.5 RECORDED DOSE PRACTICES

Recorded dose quantities at AI are given in Table 6-5 and include those provided by the site, 

Landauer (the site vendor), and special dosimeters (Hart 1979). The special dosimeters were

processed by site personnel using site calibrations, and only the results were sent to the vendor for 

inclusion in the individual’s total dose for that period. This could have resulted in different quantities, 

on occasion, from those listed in Table 6-5. 

Table 6-5. Recorded dose practices.

a

Period Dosimeter measured quantities Compliance dose quantitiesb

1954–1962 + NTA film Beta = OW, mremc

Photon (P), mR

Fast Neutron (FN), mrem

Skin = OW + P

WB = P + FN

1963–1984

Landauer + NTA film

Beta (B) or nonpenetrating, mR

Photon (P), mR

Fast Neutron (FN), mrem

Skin = (B) + P

WB = B + P + FN

WB = 0.15 (B-X-ray) + P + FNd 

1985–present Nonpenetrating (Npen)

Penetrating (Pen)

Fast Neutron (FN)

Skin = Npen + WBe

WB = Pen + neutronse 

a. OW = open window.

b. Before 1985, the Landauer assessment or calibration quantity was the roentgen at the surface of the body. In 1985, 

Landauer switched to International Commission on Radiation Units and Measurements tissue doses to incorporate 

American National Standards Institute Standard N 13.11 (HPS 1983) and DOE Laboratory Accreditation 

PerformanceProgram (DOELAP)(DOE 1986) performance testing dose quantities (Yoder 2005).

c. Garcia and Carpenter (1963 b) discusses the use of eye and gonadal shields that, if used, reduced beta values by a 

factor of 2.

d. Garcia and Carpenter (1963 b).

e. Garcia (1970).

Garcia and Carpenter (1963a ) provided a formula for correcting beta doses measured by the special 

dosimeters that is different from that in Garcia (1970). The differences are not great and are limited to 

beta dose corrections. The differences could be the result of a change from DuPont to Kodak film, 

which could have occurred in the period between the two procedures. Landauer changed to Kodak 

film in 1968 with the introduction of its Gardray film badge (Yoder 2005).
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 11 of 17

6.6 INTERPRETATION OF REPORTED DOSES 

Table 6-6 summarizes the guidance on the interpretation of reported doses. Personnel doses from 

Landauer before 1985 were reported in milliroentgen as either penetrating (photon) or nonpenetrating 

(beta) exposure if beta activity was present. The reported total values included any special dosimeter 

results sent by AI for that exchange period. Dose reconstructors should obtain organ doses, as 

needed, from External Dose Reconstruction Implementation Guideline (NIOSH 2002). 

In general, reported doses have been corrected for background using site-furnished controls. The 

controls were dosimeters at locations on the site used for storing personnel dosimeters. All personnel 

dosimeters were stored in storage racks when workers left the site at the end of their shifts. No 

workers took their dosimeters home at night; all were kept at the site and, as far as could be 

determined, this practice was in place since startup.

The exception to background correction could occur if individual dosimeters were not exchanged in 

time to be included with the regular exchange because controls were sent only with each exchange 

batch. While this could lead to an incorrect result, that result would be favorable to claimants because

it would include background and, therefore, err on the high side (Rowles 1988). 

Table 6-6. Interpretation of reported dose. 

Period

Reported 

quantity Description

Interpretation 

of zeros

Interpretation 

of blanks 

(no data)

Rollup of 

individual and 

annual dataa Monitored/ unmonitored

1954–

1960 

Skin = rad

WB = mR

Neutrons = rem

mrad, mR, and 

mrem used 

interchangeably

MDL/2 times 

number of 

zeros

If no dosimeter 

for that period, 

treat as 

unmonitored.

If special 

dosimeters were 

used, include 

results.

All personnel expected to 

be exposed to 

>100 mrem in an 

exchange period were 

monitored.

1960–

present

Skin = mrem

WB = mrem

Neutrons = mrem

mrem used for 

all

MDL/2 times 

number of 

zeros

If no dosimeter 

for that period, 

treat as 

unmonitored.

If special 

dosimeters were 

used, include 

results.

Only those >10% of 

current standard were 

monitored. Those 

entering controlled areas 

were issued a visitor 

dosimeter.

a. If special dosimeters were used, the results were forwarded to the vendor for inclusion in the totals for that period. It has not 

been determined if this was always accomplished.

6.7 ADJUSTMENTS TO RECORDED DOSE

Because most but not all penetrating photons are above 30 keV, it is suggested that an adjustment is 

necessary to account for the contribution to Hp(10) from low-energy photons from uranium and 

thorium. It is estimated that a correction equal to 10% of the <250-keV values should be made. 

However, corrections can be applied to the total WB dose by using 10% of the <250-keV value, which 

is 25%, for a total adjustment of 2.5%. Therefore, a correction factor of 1.025 should be applied to the 

total WB doses. This adjustment also increases the nonpenetrating Hp(0.07) dose, regardless of 

which dosimeter was used, because that dose was always the sum of the OW and WB doses. Table 

6-7 lists these corrections.

Table 6-7. Adjustments to recorded dose. 

Period Dosimeter Facility Adjustments to reported dose

1954–1963 Site-specific All facilities Multiply reported WB dose by 1.025 to estimate Hp(10).

1963–1985 Landauer All facilities Multiply reported WB dose by 1.025 to estimate Hp(10).

1985–present Landauer All facilities Same as above even with Landauer becoming certified by the 

National Voluntary Laboratory Accreditation Program (NVLAP).
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 12 of 17

6.8 BIAS AND UNCERTAINTY

Bias and uncertainty values were not found in any site data. The values given in Table 6-8 are from 

the Hanford TBD (ORAUT 2004) and should be similar to those at AI for the multielement dosimeter. 

These values agree with those of Landauer. In addition, AI used NTA film for its neutron dosimeter as 

did Y-12, so data from the Y-12 TBD were used to estimate the AI dosimeters bias and uncertainties

(ORAUT 2006a). The results are an estimated error of +/- 50% at 100 mrem for the 0.1- to 2-MeV 

energy range and +/- ±35% for the 2- to 14-MeV range. The latter estimate was obtained using the 

product of the ratio of the correction factors given in Table 6-4 for the two energy ranges and the +/- 

50% error. This is a reasonable assumption because most if not all the neutron energies are above 

0.5 MeV.

Table 6-8. Bias and uncertainty.

a

Site-specific 

dosimetry system

Bias magnitude and range Uncertainty factors

Overall biasb Range in biasc Systematicd Randome

Two element (1954–1963) 1.27 1.13–1.60 1.2 1.8

Multielement (1964–present)f 1.02 0.86–1.12 1.1 1.4

NTA film 2 to 14 MeV 1.5 0.50–1.50 1.5 Unknown

0.1 to 2 MeV 1.35 0.65–1.35 1.35 Unknown

a. Values are repeated here from the Hanford External Dosimetry TBD (ORAUT 2004) because the 

dosimeters in use at AI are assumed to be similar.

b. Based on the most likely distribution of energy levels and geometry. Divide recorded dose by bias value 

to determine deep dose.

c. Range of overall bias values.

d. Systematic uncertainty due to lack of knowledge of actual distributions of energies and geometries.

e. Random uncertainty due to variation among workers in energy levels and geometry.

f. These values agree with Landauer since NVLAP certification (Yoder 2005).
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 13 of 17

REFERENCES

AI (Atomics International), 1960, “Radiological Safety at Atomics International, A Division of North 

American Aviation, Inc.,” AI-MEMO 5468, July 13. [Data Capture Document 030005858] 

ANSI (American National Standards Institute), 1983, American National Standard, Criteria for Testing 

Personnel Dosimetry Performance, ANSI N13.11, New York.

Author unknown, 1956, “Examples of Data from Old Dosimetry Files.” [Data Capture Document 

0300063724]

Barnes, J. G, 2006, “Background Doses”, e-mail to L. G. Faust (M.H. Chew & Associates, Inc.),The 

Boeing Co., Canoga Park, CA, October 10.

Clow, H. E., 1966, “Neutron Survey of ’SRE Hot Cell’ Containing a 14 MeV Neutron Generator,” 

memorandum to E. T. Bramlitt, Atomics International, February 3. [Data Capture Document 

030005737] 

DOE (U. S. Department of Energy), 1986, Department of Energy Standard for the Performance 

Testing of Personnel Dosimetry Systems, DOE Laboratory Accreditation Program for 

Personnel Dosimetry Systems, DOE/EH-0027, Assistant Secretary for Environment, Safety, 

and Health, Washington, D.C., December. [SRDB Ref ID: 12294]

Etherington, H., editor, 1958, Nuclear Engineering Handbook, McGraw-Hill Book Company.

Garcia, R. R., 1970, Dose Interpretation, Reporting, and Controls in the AI Personnel Film Badge 

Program, PM-15, North American Aviation, October 30. [Data Capture Document 030005857]

Garcia, R. R., and G. D. Carpenter, 1963a, Handling of Routine Film Badges, North American 

Aviation, April 5. [Data Capture Document 030005886]

Garcia, R. R., and G. D. Carpenter, 1963b, Dose Determination, PM-14, North American Aviation, 

July 15. 

Hart, R. S., 1979, Annual Review of Radiological Controls-1975, N001T1990002, Rockwell 

International, July 2. [Data Capture Document 030005851]

HPS (Health Physics Society), 1983, An American National Standard, Personnel Dosimetry 

Performance – Criteria for Testing, ANSI/HPS N13.11-1983, McLean, Virginia.

ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England.

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Implementation Guidelines, OCAS-IG-001, Rev. 1, Office of Compensation Analysis and 

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Site – External Dosimetry, ORAUT-TKBS-0006-6, Rev. 00, Oak Ridge, Tennessee, January 9.
Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 14 of 17

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Tennessee, June 5.

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Document No. ORAUT-TKBS-0038-6 Revision No. 01 Effective Date: 11/16/2006 Page 15 of 17

GLOSSARY

accreditation

Recognition that a dosimeter system has passed the performance criteria of the DOE 

Laboratory Accreditation Program (DOELAP), (DOE 1986), in specified irradiation categories 

or the National Voluntary Laboratory Accreditation Program (NVLAP). 

accuracy

If a series of measurements has small systematic errors, they are said to have high accuracy. 

The accuracy is represented by the bias.

beta particle

A charged particle of very small mass emitted spontaneously from the nuclei of certain 

radioactive elements. Most (if not all) of the direct fission products emit (negative) beta 

particles. Physically, the beta particle is identical with an electron moving at high velocity.

favorable to claimants 

Using a process of estimation based on technical considerations such that the estimated dose 

is not underestimated.

deep dose equivalent [Hp(10)] 

The dose equivalent at the respective depth to 10 millimeters in tissue.

dose equivalent (H) 

The product of the absorbed dose (D), the quality factor (Q), and any other modifying factors. 

The special unit is the rem. When D is expressed in gray, H is in seiverts (1 sievert = 

100 rem.)

dosimeter

A device used to measure the quantity of radiation received. A holder with radiation-absorbing 

elements (filters) and an insert with radiation-sensitive elements packaged to provide a record 

of absorbed dose or dose equivalent received by an individual.

dosimetry system

A system used to assess dose equivalent from external radiation to the whole body, skin, or 

extremities. This includes the fabrication, assignment, and processing of the dosimeters as 

well as interpretation of the results.

exchange period (frequency)

Period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.

exposure

A measure expressed in roentgens of the ionization produced by gamma (or X) rays in air.

film

Generally means a “film packet” that contains one or more pieces of film in a light-tight 

wrapping. The film when developed has an image caused by radiation that can be measured 

using an optical densitometer.

film dosimeter

A small packet of film within a holder that attaches to a worker.
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filter

Material used to adjust radiation response of a dosimeter to provide an improved tissue 

equivalent or dose response.

gamma rays

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

nuclear reactions (e.g. fission, radioactive decay, and neutron capture). Physically, gamma 

rays are identical to X-rays of high energy, the only essential difference being that X-rays do 

not originate in the nucleus.

ionizing radiation

Electromagnetic radiation (consisting of photons or particulate radiation consisting of 

electrons, neutrons, protons, etc.) capable of producing charged particles through interactions 

with matter.

neutron

A basic particle that is electrically neutral weighing nearly the same as the hydrogen atom.

nonpenetrating dose [Hp(0.07)] 

Designation (i.e., NP or NPen) on film dosimeter reports that implies a radiation dose, typically 

to the skin of the whole body, from beta and lower energy photon radiation.

open window (OW) 

Designation on a dosimeter that implies the use of little or no shielding. It commonly is used to 

label the film response corresponding to the open window area.

penetrating dose [Hp(10)] 

Designation (i.e. P, Pen, or Neutrons and Gamma) on a dosimeter of the dose recorded at 

depth of 10 millimeters.

personal dose equivalent [Hp(d)] 

Radiation quantity recommended for use as the operational quantity to be recorded for 

radiological protection purposes by the International Commission on Radiological Units and 

Measurements. The personal dose equivalent is represented by Hp(d), where d identifies the 

depth in millimeters and represents the point of reference for dose in tissue. For weakly 

penetrating radiation of significance to skin dose, d = 0.07 millimeter and is noted as Hp(0.07). 

For penetrating radiation of significance to whole-body dose, d = 10 millimeters and is noted 

as Hp(10). 

photon

A unit of electromagnetic radiation consisting of X- and/or gamma rays.

PM

A procedure detailing specific actions or directions and usually limited to one service or 

activity.

rad

A unit of absorbed dose equal to the absorption of 100 ergs per gram of absorbing material 

such as body tissue.
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radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

rem

A unit of dose equivalent in human tissue equal to the product of the number of rads and the 

quality factor and any other modifying factors. The word derives from roentgen equivalent in 

man. 

rep

Historically, used extensively for the specification of permissible doses of ionizing radiations 

other that X-rays or gamma rays. Several definitions have appeared in the literature but the 

most widely adopted is a unit of absorbed dose with a magnitude equal to 93 ergs per gram. 

The word derives from roentgen equivalent physical. 

roentgen

A unit of exposure to gamma or X-rays. It is defined precisely as the quantity of gamma (or X) 

rays that will produce a total charge of 2.58 × 10-4 coulomb in 1 kilogram of dry air. An 

exposure of 1 roentgen is approximately equivalent to an absorbed dose of 1 rad in soft tissue.

sievert

The International System unit for dose equivalent (1 sievert = 100 rem).

skin dose

Absorbed dose at a tissue depth of 0.07 millimeters (7 milligrams per square centimeter). 

tissue equivalent

Term used to imply that the radiation response of the material being irradiated is equivalent to 

tissue.

whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 10 millimeters (1,000 milligrams

per square centimeter); also used to refer to the “dose of record.” 

X-ray

Ionizing electromagnetic radiation of extra-nuclear origin.